Radioactive waste is generated primarily by nuclear power plants but also by hospitals, research laboratories and the like. Radioactive waste in the United States falls into a number of categories; most prominent are High Level Waste (HLW) and Low Level Waste (LLW). Included in the category of HLW is spent nuclear fuel along with some other radioactive waste not considered suitable for near surface disposal. LLW is generated at and within the controlled areas of commercial nuclear reactors, in hospitals, at public and private research facilities, and the like. Of particular concern is the LLW generated at nuclear power plants relative to the filtration processes employed to keep the water surrounding the reactor and the spent fuel clean and free of contaminating ions. Such filter material is comprised of such things as ion exchange resins, filter media and cartridge filters.
Primary responsibility for the disposition of HLW and federally generated LLW falls under the United States Department of Energy (DOE). Responsibility for the disposition of commercially generated LLW (e.g. hospitals, institutions, industrial generators, as well as nuclear power plants) is assigned to the States by the Low Level Waste Policy Act of 1980, as amended in 1985. The Nuclear Regulatory Commission (NRC) provides general criteria for the disposal of LLW through Title 10, Part 61 of the Code of Federal Regulations. Section 61.55 thereof defines three classes of LLW; A, B and C. Class A waste is the least radioactive, B is intermediate and C is the most radioactive. Class A is suitable for near ground level storage, whereas B and C level waste require progressively more secure and deeper storage which entails progressively greater storage expense. Classification depends upon the presence or absence of certain radionuclides and the concentrations thereof. Those regulated radionuclides are set out in tables one and two below.
TABLE 1Radionuclides from 10 CFR 61.55 Table 1ConcentrationTable 1 Radionuclidecuries per cubic meterC-148C-14 in activated metal80Ni-59 in activated metal220Nb-94 in activated metal0.2Tc-993I-1290.08Alpha emitting transuranic nuclides with1100half-life greater than 5 yearsPu-24113,500Cm-242120,0001Units are nanocuries per gram.
TABLE 2Radionuclides from 10 CFR 61.55 Table 2Concentration, curiesper cubic meterTable 2 RadionuclideCol. 1Col. 2Col. 3Total of all nuclides with less than 5 year700(1)(1)half-lifeH-340(1)(1)Co-60700(1)(1)Ni-633.570700Ni-63 in activated metal357007000Sr-900.041507000Cs-1371444600(1) There are no limits established for these radionuclides in Class B or C wastes.
For Table 1 radionuclides, if the concentration is less than the table limit, the waste is Class C. If the concentration is less than 0.1 times the table limit the waste is Class A. For Table 2 radionuclides, each column represents the Class limit. For mixtures of radionuclides, the sum of the fractions determines the waste class as described in 10 CFR 61.55. In these tables, only 60Co and 137Cs are easily detected via gamma spectroscopy. All other radionuclides require detailed laboratory analysis and/or chemical separation to be quantified directly.
Generators of LLW are required to report the quantity of radioactivity associated with specific radionuclides to the disposal sites that are licensed to accept this waste. These specific radionuclides of importance for disposal are identified in the Code of Federal Regulations (CFR), Title 10; Part 61. This identification is part of an overall process to describe the physical, chemical, and radiological characteristics of the waste and to verify its suitability within various legal requirements for ultimate disposal. The required quantification of radioactivity is referred to as radiological characterization. The disposal site operators rely on this information to verify that the waste meets specific disposal requirements and to track overall inventory within allowable gross activity limits applied to the disposal site.
Radiological characterization involves determining what radionuclides are present in the waste and their concentrations so they can be disposed of properly. Such measurements are conducted in accordance with NRC guidance provided in two LLW Branch Technical Position (BTP) documents. The first NRC BTP document entitled “Waste Classification” issued in 1983 identifies four methods for determining radioactivity in waste. These include:
Materials Accountability—maintaining a specific inventory of radionuclides as received or created and adjusted for disposal.
Classification By Source—understanding and tracking the origin of the materials along with their radioactivity content.
Gross Radioactivity Measurements—Using survey instruments to measure radioactivity by energy absorption (dose rate) and inferring the concentrations from data collected from material of the same or comparable origin. This approach however cannot be used in isolation since prior knowledge of the relative distribution of radionuclides is required.
Direct measurement of individual radionuclides—Sampling and analyzing the specific material to be disposed of.
Any or all of these methods can be used to characterize the class of the LLW. The second NRC BTP document on “Concentration Averaging and Encapsulation” issued in 1995 set limitations on the ability to average wastes of different radioactive isotopic concentrations and gave guidance for determining weights and volumes to be used for classification. The 1995 BTP did not limit any of the methods that could be used to initially determine activity content. The methods currently used tend to rely primarily on gross radioactivity measurements and direct measurement methods. The largest volumes of radioactive waste are generated by nuclear power plants and are most often characterized using single samples of the final waste form. Difficult-to-measure radionuclides are measured either directly on the basis of a single sample or using statistical correlations to radionuclides that are more readily measured with available equipment based on sample data collected over 3-5 years.
However, the characterization process of using samples of the final waste form is flawed for a number of reasons. Obtaining truly representative samples of utility waste streams is problematic. Since regulatory guidance didn't come into play until after the final designs of today's operating plants, many waste characterization and processing systems do not have the capability of sampling until wastes from multiple collection points have been transferred to a collection/receiving container or tank. Some systems do not have sampling capabilities until after the waste has been transferred into the actual disposal containers. Mixing capabilities are varied with some sites having no mixing capability at all such that the waste material has little or no homogeneity. Tank designs or operations may not allow complete removal of all material on discharge to a waste container resulting in a mixture of waste from different operating periods. The collection of multiple samples for a particular batch of waste is frequently not possible due to worker radiation exposure concerns and waste sample sizes are frequently limited to less than 500 grams. The size of a sample sent to an offsite laboratory is often limited by activity and laboratory license restrictions to amounts less than 1 gram. On-site samples used for gamma spectroscopy are sometimes less than 1 milligram. The results from such samples cannot be assumed to be truly statistically representative of all the material in a particular storage container. As is well understood, the number of samples (n) required to establish a statistically meaningful result at a specified precision can be calculated using the following equation:n=ta2+S2/D2  Equation 1Where, D represents the precision level,ta is the Student T-test Value, andS2 is the standard deviation of the data set.
Based on the foregoing equation, to achieve a precision of a factor of 10, as required by the previously referenced BTP on concentration averaging, requires a sample set S2=0.628. The result of this calculation is that 116 samples are needed to achieve the required precision. At a cost of approximately $2,000 per sample analysis, increasing the number of samples to this level is not a viable approach from an overall cost perspective.
The derivation of scaling factors for the difficult to detect radionuclides using common laboratory practices is also fraught with inaccuracy. Following current practice, the inventory of 129I, for example, would be determined by a scaling factor derived from endpoint sampling. As seen by referring to the scaling factor plot of FIG. 1, represented therein is scaling factor data from a nuclear power plant based on end-point waste samples. In this case, the quantity of 129I is shown as a ratio to the quantity of 137Cs. Since 137Cs is more easily detected, the quantity of 129I in the waste would be determined by multiplying the 137Cs activity by the scaling factor.
In this case, the scaling factor based on detection limits would be <4.36×10−4 for primary resins alone. A corresponding average 137Cs value is 0.961 μCi/cc. Scaling up from these values results in an estimated 137Cs inventory of ˜0.091 Curies and a corresponding 129I value of (something less than) 39.56 μCi. This compares with the coolant based estimated total release of 0.0514 μCi over the same time period. The reported activity from scaling factors derived from waste sampling is almost 3 orders of magnitude higher. Current practices provide no alternative to reporting the most restrictive value. Another point to be noted from the data above is that the five values indicated by asterisks in FIG. 1 by are not actual quantities for 129I but are in fact the detection limits of the analysis technique. The actual value is therefore something less than the reported value. Also, the plot indicates that for a change in 137Cs values over approximately three orders of magnitude, the 129I values remain the same. There is therefore, no correlation between the two radionuclides and the scaling factor derived from this data is extremely inaccurate at best.
Thus, standard practice is to err on the conservative side and categorize a LLW as being in a more restrictive class such as Class B instead of Class A, or Class C instead of B. Additionally, even where evaluating data from multiple samples is done, prior art characterization procedure involves preferentially including high-side variations in each of the sample values and excluding low side variations in an attempt to avoid the perception of a non-conservative evaluation. In both cases, the result is to frequently over specify concentrations of disposal limiting radionuclides which often leads to significantly higher disposal costs and a waste of disposal resources than would otherwise be required if the waste was more accurately characterized. More precise characterization of such waste could save nuclear power plants millions of dollars in yearly operating costs. In addition, determining waste classification and disposal options after the waste is generated precludes the ability of the generator to manage the costs of waste generation and accurately predict operating costs.